The number of light-water reactors beyond 40 years of operation has increased, raising awareness of the problem of degradation over time of structural members. One type of degradation over time is stress-corrosion cracking (hereinafter referred to as SCC). SCC occurs when three factors, i.e. material, environment and stress, act simultaneously.
In the pressure boundary of a light-water reactor, Alloy 600 (15Cr-70Ni—Fe) or Alloy 690 (30Cr-60Ni—Fe) are used at positions that require particularly good SCC resistance. Alloy 690 has been commercialized as a material that improves Alloy 600 in terms of SCC initiation, where one of its features is that it has been subjected to a special heat treatment that intentionally precipitates M23C6 on grain boundaries and resolves Cr-depleted layers.
An example special heat treatment is described in Yonezawa et al., “Effects of Metallurgical Factors on Stress Corrosion Cracking of Ni-Base Alloys in High Temperature Water”, Proceedings of JAIF International Conference on Water Chemistry in Nuclear Power Plants, volume 2 (1988), pp. 490-495.
Various methods to improve the SCC resistance of these alloys have been disclosed. Japanese Patent No. 2554048 discloses a high-strength Ni-based alloy having at least one of a γ′ phase and γ″ phase in the γ base and providing a microstructure in which M23C6 has precipitated with priority in a semi-continuous manner on crystal grain boundaries to improve SCC resistance. Japanese Patent No. 1329632 and JP Sho60 (1985)-245773 A each disclose an Ni-based alloy where a heating temperature and a heating time after cold rolling are specified to improve SCC resistance. Japanese Patent No. 4433230 discloses a high-strength Ni-based alloy pipe or tube for nuclear power where the crystal grain size is made fine by Ti- or Nb-containing carbonitrides.